ML20033B539
| ML20033B539 | |
| Person / Time | |
|---|---|
| Site: | Atlantic Nuclear Power Plant  |
| Issue date: | 11/27/1981 |
| From: | Lauben G, Powers D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20033B480 | List: |
| References | |
| NUDOCS 8112010504 | |
| Download: ML20033B539 (6) | |
Text
~
N pl7 7 i
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMI11SSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
0FFSHORE POWER SYSTEMS Docket No. STN 50-437 (Floating Nuclea.- Power Plants)
)
NRC STAFF:S UPDATE OF ITS ECCS ANALYSIS FOR THE FHP by G. Noman Lauben Dale A. Powers The purpose of this testimony is to address the impact of ECCS analysis of the current Staff position on fuel cladding swelling and rupture.
In reference 1, this Board was notified of the Staff determination that:
[P] arts of the approved ECCS evaluation models, for all light water reactors, relating to cladding rupture temperature, steam and flow blockage might, in general, be non-conservative and there-fore, might not be in compliance with Appendix K of 10 C.F.R. 50. Section IB of Appendix K states:
"To be acceptable the swelling 2nd rupture calculations shall be based upon applicable data in such a way that the degree of swelling and incidence of rupture are not underestimated."
The Staff indicated that further information would be forthcoming as it became available. The inforTnation provided in this testimony addresses this issue in the same manner as has been done for all Westinghouse designed reactors in operation or being reviewed for an operating license in the near term.
8112010504 d11127 POR ADOCK 05000437 PDR
O g
- The NRC Staff has been generically evaluating three materials models that are used in ECCS evaluations. Those models predict cladding rupture temperature, cladding burst strain, and fuel assembly flow blockage. We have (a) discussed our evaluation with vendors and other industry representatives (Reference 2), (b) published NUREG-0630,
" Cladding Swelling and Rupture liodels for LOCA Analysis" (Reference 3),
and (c) required licensees to confirm that their operating reactors would continue to be in conformance with 10 C.F.R. 50.46 if the NUREG-0630 models were substituted for the present materials models in their ECCS evaluations and certain other compensatory model changes were allowed (Reference 4 and 5).
Until we have completed our generic review and implemented new acceptance criteria for cladding models, we have required that the ECCS analyses be accompanied by supplemental calculations to be performed with the materials models of NUREG-0630. For these supplemental calculations, we have accepted other compensatory model changes that have been approved by the NRC, but have not been utilized fonnally by the Applicant. These changes are consistent with the changes allowed for the confirmatory operating reactor calculations mentioned above.
In Reference 6, the Applicant provided the current ECCS analysis of reconi for the Floating Nuclear Plant.
In that report it is stated that the analysis was performed for a nearly identical reference plant.
In Reference 7, the Applicant identified the reference plant as the licGuire Nuclear Station.
In Reference 8, Duke Power Co. submitted the supplemental calculation discussed above for the licGuire plant.
In Reference 7, the Applicant adopted the McGuire supplemental calculation
as applicable to the Floating Nuclear Plant. The Staff has reviewed and approved the McGuire submittal and we agree that for the purposes of this assessment that is an appropriate method of reference.
Reference 8 also addressed a recently identified non-conservatism of the Westinghouse 1978 ECCS evaluation model. The new concern was discovered by Westinghouse who formally notified the Staff in November 1979 (Reference 9). Specifically, Westinghouse had discovered 'that the February 1978 ECCS evaluation model was, in part, based on cladding burst tests which were conducted at relatively fast temperature-ramp rates; whereas the LOCA analyses of actual plant heatup rates (including those of FNP) were at relatively slow temperature-ramp rates. Reference 8 assessed the combined impact of this calculational error and the NUREG-0630 models to be worth 855 F peak cladding temperature above that previously calculated.
Subsequently Westinghouse calculated that a reduction in total peaking factor, F, of 0.025 would offset the increase q
in peak cladding temperature above 2200*F. However, Westinghouse identified a margin in F available through the use of a reduction in q
pellet temperature uncertainty (see Reference 10 for approval). This margin was worth 0.031 in F. Thus no F reduction was required.
q q
We therefore, conclude that the Applicant has satisfied cJr concerns related to the swelling and rupture issue.
. -.,,. - - -, _ -.. - ~., _, - - _.
REFERElCES 1.
Board Notification:
Rupture Strain and Flow Blockage Models in PWR LOCA Calculations, BN-79-36 dated Nov. 2,1979.
2.
tiemorandum from R. P. Denise, NRC, to R. J. Mattson, " Summary Minutes of Meeting on Cladding Rupture Temperature, Cladding Strain, and Assembly Flow Blockage," November 20, 1979.
3.
D. A. Powers and R. O. Meyer, " Cladding Swelling and Rupture Models for LOCA Analysis," NRC Report NUREG-0630, April 1980.
4.
Letter from D. G. Eisenhut, NRC, to all Operating Light Water Reactors, dated November 9, 1979.
5.
Memorandum from H. R. Denton, HRC, tr. Convaissioners, " Potential Deficiencies in ECCS Evaluation Models," November 26, 1979.
6.
UHI/ECCS Analysis for the Floating Nuclear Plant, Topical Report No. 24A37, Supplement No.1.
7.
Letter from P. B. Haga, Offshore Power System, to Elinor Adensam, NRC, on ECCS/VHI Perfonnance Evaluation dated Nov. 11, 1981.
8.
Letter from illiam 0. Parker, Jr., Duke Power Company, to B. J.
Youngblood, NRC, dated September 15, 1980.
9.
Letter from T. M. Anderson, Westinghouse Electric Corporation, to D. G. Eisenhut, NRC, number NS-TMA-2163, dated November 16, 1979.
- 10. Letter from J. F. Stolz, NRC, to T. M. Anderson, Westinghouse Electric Corporation, Review of WCAP-8720, " Improved Analytical flodels Used in Westinghouse Fuel Rod Design Computations," dated flarch 27, 1980.
e STATEMTENT OF PROFESSIONAL QUALIFICATIONS NORMAN LAUBEN liy name is George Norman Lauban.
I am emplrayed as a Nuclear Engineer in the Reactor Systems Branch, Division of Systems Integration, U.S. Nuclear Regulatory Commission.
I have worked 'in the field of nuclear reactor safety for 19 years, and in nuclear activities for 23 years.
I have worked for the Commission and its predecessor, the Atomic Energy Commission, since 1968.
During this time I have worked directly on reactor safety matters, including Emergency Core Cooling System (ECCS) performance review and Loss-of-Coolant Accident (LOCA) analysis.
I was a member of the 1971 AEC ECCS task force and the AEC Staff Panel for the ECCS Rulemaking H2aring.
I am the author of the T00DEE2 computer program used by the NRC and the nuclear industry for transient fuci pin thermal analysis during a LOCA.
I was a member of the technical team that accompanied Mr. Harold Denton to the Three Mile Island Reactor on March 30, 1979.
I have a B.S. and M.S. in Chemical Engineering from Case Institute of Technology (now Case Western Reserve University).
PROFESSIONAL QUALIFICATIONS OF DALE A. POWERS U.S. Nuclear Regulatory Commission Washington, D. C.
I am employed as the Senior Reactor Fuels Metallurgist of the Core Performance Branch, Office of Nuclear Reactor Regulation. As such, I am responsible for the review and evaluation of the metallurgical design and performance of nuclear power plant fuel systems and components.
Areas of detailed technical review include those portions of applicant and licensee safety analyses for CP, OL, and reload applications, which address such issues as fuel rod bowing, cladding creep, pellet / cladding interactions, etc.
I have served in this capacity since 1976.
My general background is that of a nuclear and metallurgical engineer with primary experience in the analysis of Zircaloy fuel cladding during both LWR normal operating and accident conditions, ultrasonic methods of assessing material properties in radiation-damaged aluminum alloys, radio-tracer techniques for determining interstitial diffusion and phase transformations in iron alloys, and testing procedures for nuclear power plants.
From 1970 to 1971 I was employed as a Shift Test Engineer in the Nuclear Power Division at Mare Island Naval Shipyard, where I was responsible for the preparation and direction of on-board testing operations associ-ated with the construction and overhaul of naval nuclear propulsion plants.
The specific areas of my involvement included pre-operational flushing and hydrostatic testing of primary and secondary coolant systems, t ectrical systems interaction testing, and initial startup (criticality and power range) testing.
I completed the Naval Nuclear Power School at General Dynamics.
I have received a B.S. degree (1970) in Metallurgical Engineering with a Nuclear Option, an M.S. degree (1974) in Nuclear Engineering, and a Ph.D degree (1977) in Metallurgical Engineering from the University of Missouri - Rolla.
During graduate schoo', I held research and teaching assistantships in physical metallurgy and reactor physics, respectively.
I am a member of the Honor Societies of Phi Kappa Phi, Alpha Sigma Mu, and Nuclear Engineering and Science.
I have authored technical papers for presentation at conferences or publication (including NUREG-0630,
" Cladding Swelling and Rupture Models fcr LOCA Analysis").
b